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Journal Articles

Experimental study on local damage to reinforced concrete panels subjected to oblique impact by projectiles

Okuda, Yukihiko; Nishida, Akemi; Kang, Z.; Tsubota, Haruji; Li, Y.

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021801_1 - 021801_12, 2023/04

Most empirical formulas were proposed to evaluate the local damage to reinforced concrete (RC) structures based on impact tests conducted with a rigid projectile at an impact angle normal to the target structure. Only a few impact tests were performed involving a soft projectile. Therefore, in this study, we conducted a series of impact tests to evaluate the local damage to RC panels subjected to normal and oblique impacts by rigid and soft projectiles. This paper presents the test conditions, test equipment, test results, and obtained knowledge on local damage to RC panels subjected to normal and oblique impacts.

JAEA Reports

Guideline on structural integrity assessment for reactor pressure vessel in domestic light water reactor based on probabilistic fracture mechanics

Lu, K.; Katsuyama, Jinya; Takamizawa, Hisashi; Li, Y.

JAEA-Research 2022-012, 39 Pages, 2023/02


For reactor pressure vessels (RPVs) in the light water reactors, the fracture toughness decreases due to the neutron irradiation embrittlement with operating years. In Japan, to prevent RPVs from a nil-ductile fracture, deterministic fracture mechanics methods in accordance with the codes provided by the Japan Electric Association are performed for assessing the structural integrity of RPVs under the pressurized thermal shock (PTS) events by taking the neutron irradiation embrittlement into account. On the other hand, in recent years, probabilistic methodologies for PTS evaluation are introduced into regulations in the United States and some European countries. For example, in the United States, a PTS screening criterion related to the reference temperature based on the probabilistic method is stipulated. If the screening criterion is not satisfied, it is allowable to perform the evaluation based on the probabilistic method by calculating numerical index such as through-wall crack frequency (TWCF). In addition, the reduction of non-destructive examination extent or extension of examination intervals for RPV welds have been discussed based on the probabilistic method. Here, the probabilistic method is a structural integrity assessment method based on probabilistic fracture mechanics (PFM) which is rational in calculating the failure probability of components by considering uncertainties of various factors related to the aged degradation due to the long-term operation. Based on these backgrounds, we developed a PFM analysis code PASCAL and released a guideline on structural integrity assessment based on PFM by reflecting the latest knowledge and expertise in 2017. Here, the main analysis target was the RPV of pressurized water rector considering neutron irradiation embrittlement and PTS events in the structural integrity assessment of RPVs. The objective of the guideline is that persons who have knowledge on the fracture mechanics can carry out the PFM analyses and

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.5 for reactor pressure vessels

Takamizawa, Hisashi; Lu, K.; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.

JAEA-Data/Code 2022-006, 221 Pages, 2023/02


As a part of the structural integrity assessment research for aging light water reactor (LWR) components, a probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in Japan Atomic Energy Agency. The PASCAL code can evaluate failure probabilities and failure frequencies of core region in reactor pressure vessel (RPV) under transients by considering the uncertainties of influential parameters. The continuous development of the code aims to improve the reliability by introducing the analysis methodologies and functions base on the state-of-the-art knowledge in fracture mechanics and domestic data. In the first version of PASCAL, which was released in FY2000, the basic framework was developed for analyzing failure probabilities considering pressurized thermal shock events for RPVs in pressurized water reactors (PWRs). In PASCAL Ver. 2 released in FY 2006, analysis functions including the evaluation methods for embedded cracks and crack detection probability models for inspection were introduced. In PASCAL Ver. 3 released in FY 2010, functions considering weld-overlay cladding on the inner surface of RPV were introduced. In PASCAL Ver. 4 released in FY 2017, we improved several functions such as the stress intensity factor solutions, probabilistic fracture toughness evaluation models, and confidence level evaluation function by considering epistemic and aleatory uncertainties related to influential parameters. In addition, the probabilistic calculation method was also improved to speed up the failure probability calculations. To strengthen the practical applications of PFM methodology in Japan, PASCAL code has been improved since FY 2018 to enable PFM analyses of RPVs subjected to a broad range of transients corresponding to both PWRs and boiling water reactors, including pressurized thermal shock, low-temperature over pressure, and normal operational transients. In particular, the stress intensi

Journal Articles

Fatigue crack non-propagation behavior of a gradient steel structure from induction hardened railway axles

Zhang, H.*; Wu, S. C.*; Ao, N.*; Zhang, J. W.*; Li, H.*; Zhou, L.*; Xu, P. G.; Su, Y. H.

International Journal of Fatigue, 166, p.107296_1 - 107296_11, 2023/01

 Times Cited Count:0 Percentile:0.03(Engineering, Mechanical)

Journal Articles

Uncertainty quantification of seismic response of nuclear reactor building using a three-dimensional finite element model

Choi, B.; Nishida, Akemi; Li, Y.; Takada, Tsuyoshi

Earthquake Engineering and Resilience (Internet), 1(4), p.427 - 439, 2022/12

no abstracts in English

Journal Articles

Closed-form stress intensity factor solutions for circumferential and axial surface cracks with large aspect ratios in pipes

Azuma, Kisaburo*; Li, Y.

Journal of Pressure Vessel Technology, 144(6), p.061303_1 - 061303_13, 2022/12

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

Journal Articles

Stability of allowable flaw angles for high toughness ductile pipes subjected to bending stress in the ASME Code Section XI

Hasegawa, Kunio; Strnadel, B.*; Li, Y.; Lacroix, V.*

Journal of Pressure Vessel Technology, 144(6), p.061202_1 - 061202_6, 2022/12

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

When pipe walls are thin, part-through flaws are easily develop into through-wall flaws, and the likelihood of coolant leakage is high. The ASEM Code Section XI provides final allowable flaw angles of through-wall flaw for thin-wall pipes. The final allowable angles are applied to pipes in order to maintain structural integrity if the part-through flaws become through-wall flaws. To ensure that this stability is compromised, plastic collapse stresses for through-wall flaws are combined with allowable stresses. However, the final allowable angles of through-wall flaws are not identified for thin-walled pipes. This paper compares plastic collapse stresses of through-wall flaws and allowable stresses of part-through flaws for pipes. The comparison of these stresses is used to derive the final allowable angles of through-wall flaws. The angles can be expressed either in the form of exact solutions or as conventional options that are appropriate for various service level conditions.

Journal Articles

New $$K$$ isomers in $$^{248}$$Cf

Orlandi, R.; Makii, Hiroyuki; Nishio, Katsuhisa; Hirose, Kentaro; Asai, Masato; Tsukada, Kazuaki; Sato, Tetsuya; Ito, Yuta; Suzaki, Fumi; Nagame, Yuichiro*; et al.

Physical Review C, 106(6), p.064301_1 - 064301_11, 2022/12

 Times Cited Count:0 Percentile:0.02(Physics, Nuclear)

Journal Articles

"Southwestern" boundary of the $$N = 40$$ island of inversion; First study of low-lying bound excited states in $$^{59}$$V and $$^{61}$$V

Elekes, Z.*; Juh$'a$sz, M. M.*; Sohler, D.*; Sieja, K.*; Yoshida, Kazuki; Ogata, Kazuyuki*; Doornenbal, P.*; Obertelli, A.*; Achouri, N. L.*; Baba, Hidetada*; et al.

Physical Review C, 106(6), p.064321_1 - 064321_10, 2022/12

 Times Cited Count:0

The low-lying level structure of $$^{59}$$V and $$^{61}$$V was investigated for the first time. The neutron knockout reaction and inelastic proton scattering were applied for $$^{61}$$V while the neutron knock-out reaction provided the data for $$^{59}$$V. Four and five new transitions were determined for $$^{59}$$V and $$^{61}$$V, respectively. Based on the comparison to our shell-model calculations using the Lenzi-Nowacki-Poves-Sieja (LNPS) interaction, three of the observed $$gamma$$ rays for each isotope could be placed in the level scheme and assigned to the decay of the first 11/2$$^{-}$$ and 9/2$$^{-}$$ levels. The ($$p$$,$$p'$$) excitation cross sections for $$^{61}$$V were analyzed by the coupled-channels formalism assuming quadrupole plus hexadecapole deformations. Due to the role of the hexadecapole deformation, $$^{61}$$V could not be unambiguously placed on the island of inversion.

Journal Articles

Extended $$p_{3/2}$$ neutron orbital and the $$N = 32$$ shell closure in $$^{52}$$Ca

Enciu, M.*; Liu, H. N.*; Obertelli, A.*; Doornenbal, P.*; Nowacki, F.*; Ogata, Kazuyuki*; Poves, A.*; Yoshida, Kazuki; Achouri, N. L.*; Baba, Hidetada*; et al.

Physical Review Letters, 129(26), p.262501_1 - 262501_7, 2022/12

 Times Cited Count:0 Percentile:0(Physics, Multidisciplinary)

The one-neutron knockout from $$^{52}$$Ca was performed at $$sim$$230 MeV/nucleon combined with prompt $$gamma$$ spectroscopy. The momentum distributions corresponding to the removal of $$1f_{7/2}$$ and $$2p_{3/2}$$ neutrons were measured. The cross sections are consistent with a shell closure at the neutron number $$N = 32$$, found as strong as at $$N = 28$$ and $$N = 34$$ in Ca isotopes from the same observables. The analysis of the momentum distributions leads to a difference of the root-mean-square radii of the neutron $$1_{f7/2}$$ and $$2p_{3/2}$$ orbitals of 0.61(23) fm, in agreement with the modified-shell-model prediction of 0.7 fm suggesting that the large root-mean-square radius of the $$2p_{3/2}$$ orbital in neutron-rich Ca isotopes is responsible for the unexpected linear increase of the charge radius with the neutron number.

Journal Articles

Improvement of the return mapping algorithm based on the implicit function theorem with application to ductile fracture analysis using the GTN model

Mano, Akihiro; Imai, Ryuta*; Miyamoto, Yuhei*; Lu, K.; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104700_1 - 104700_13, 2022/10

 Times Cited Count:0 Percentile:0(Engineering, Multidisciplinary)

Elastic-plastic analyses based on finite element methods are widely applied to simulate the nonlinear behaviors of materials. When the analysis is conducted by an implicit method, the stress values are generally updated with a time increment by using the so-called return mapping algorithm. This algorithm requires solving simultaneous nonlinear equations related to a constitutive model. In the present paper, we proposed a general method to reduce the number of equations in the return mapping algorithm based on the implicit function theorem. In addition, the proposed method was applied to the Gurson-Tvergaard-Needleman (GTN) model that considers the influence of damage due to nucleation and growth of microscopic void in materials in the simulation of the nonlinear behaviors. By using the GTN model with the proposed method, an elastic-plastic analysis was performed by the implicit method for a 4-point bending test of pipe with a through-wall crack. The numerical solution of the variation of the load-load line displacement from the analysis agreed with experimental result. Thus, we concluded that the proposed method is useful for simulating nonlinear behaviors, including void nucleation and growth in materials.

Journal Articles

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10

 Times Cited Count:1 Percentile:65.32(Engineering, Multidisciplinary)

Journal Articles

Allowable cracks related to penetration for part-through cracks in pipes subjected to bending stresses

Hasegawa, Kunio; Li, Y.; Strnadel, B.*; Udyawar, A.*

Journal of Pressure Vessel Technology, 144(5), p.051305_1 - 051305_6, 2022/10

 Times Cited Count:1 Percentile:61.31(Engineering, Mechanical)

Fully plastic collapse stresses for circumferentially part-through cracked pipes subjected to bending stresses are estimated by Limit Load Criteria provided by the ASME Code Section XI. Allowable crack depths were determined by using the Limit Load Criteria and that are tabulated in the ASME Code Section XI for different plant service level conditions. On the other hand, crack penetration bending stresses for part-through cracked pipes were estimated by using the Local Approach of Limit Load Criteria. By using these Criteria, the study presented in this paper obtained allowable crack depths at penetration for circumferentially part-through cracked pipes. Comparing the allowable crack depths obtained by both methods for each service level, it is evident that the allowable crack depths at penetration calculated by the Local Approach of Limit Load Criteria are almost always smaller than those at fully plastic collapse stresses calculated by the Limit Load Criteria. It was found that the allowable crack depths provided by the ASME Code Section XI are less conservative for crack penetrations.

Journal Articles

Local damage to reinforced concrete panels subjected to oblique impact by projectiles; Numerical analysis on test results

Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Li, Y.

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08

Most of the empirical formulas that have been proposed seeking to quantitatively investigate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived based on impact tests performed normal to the target structure, while only a few impact tests involving soft projectile to the target structure have been studied. The purpose of this study is to develop a local damage evaluation method that takes into account the oblique impact due to soft projectile, which should be considered in realistic impact conditions. In this paper, we compare the test results with the analytical results to examine and validate the parameter setting of analytical method for evaluating local damage in RC panel. The obtained knowledge is presented.

Journal Articles

Analytical study for low ground contact ratio of buildings due to the basemat uplift using a three-dimensional finite element model

Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08

In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.

Journal Articles

Empirical correction factor to estimate the plastic collapse bending moment of pipes with circumferential surface flaw

Lacroix, V.*; Hasegawa, Kunio; Li, Y.; Yamaguchi, Yoshihito

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07

Journal Articles

Failure bending stresses for small diameter thick-wall pipes

Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.; Lacroix, V.*

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 4 Pages, 2022/07

Journal Articles

Development of probabilistic analysis code for evaluating seismic fragility of aged pipes with wall-thinning

Yamaguchi, Yoshihito; Nishida, Akemi; Li, Y.

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07

The wall-thinning is one of the most important age-related degradation phenomena in nuclear piping systems. Furthermore, in recent years, several nuclear power plants in Japan have experienced severe earthquakes. Therefore, failure probability analysis and fragility evaluation of piping systems, taking both wall-thinning and seismic response stresses into consideration, have become increasingly important in seismic probabilistic risk assessment. In Japan Atomic Energy Agency, in order to evaluate the failure probability of aged piping system with wall-thinning, a probabilistic analysis code PASCAL-EC was developed. In this study, to evaluate the seismic fragility of a wall-thinned pipe, a model of seismic response stress considering the wall-thinning effect, a failure evaluation method for wall-thinned pipes, and functions related to uncertainties treatment for important influence parameters have been introduced to PASCAL-EC. In this paper, the improved PASCAL-EC is outlined and preliminary results of the seismic fragility evaluation performed using this code are provided.

Journal Articles

Failure estimation methods for steam generator tubes with wall-thinning or crack

Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07

The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.

Journal Articles

Study on influence evaluation of internal equipment installed in structure subjected to projectile impact

Okuda, Yukihiko; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 10 Pages, 2022/07

When a projectile collides with a nuclear building, stress waves are generated at the impacted area and propagate to the interior of the building through the building structure. Assessing the influence of dynamic responses generated by the projectile impact on internal equipment is important, because stress waves are likely to excite high-frequency vibrations of the internal equipment and may influence the functionality of the internal equipment. Therefore, we performed a projectile impact test on a reinforced concrete (RC) structure that models a nuclear building with internal equipment. This paper presents the results of the investigation of the impact response characteristics of the RC structure subjected to projectile impact.

427 (Records 1-20 displayed on this page)